1. Field of the Invention
The present invention relates to zirconium-based alloys suitable for use in nuclear reactor service, and more specifically for use in the cladding of fuel elements.
2. Description of Related Art
Zirconium-based alloys have long been used in the cladding of fuel elements in nuclear reactors. A desirable combination is found in zirconium by virtue of its low thermal neutron cross-section and its generally acceptable level of resistance to corrosion in a boiling water reactor environment. Zircaloy 2, a Zr--Sn--Ni--Fe--Cr alloy, has enjoyed widespread use and continues to be used at present in nuclear reactor applications. This alloy has provided adequate performance in reactor service, but also possesses some deficiencies which have prompted further research to find materials which would provide improved performance. Zircaloy 4 was one alloy developed as a result of that research. This allloy essentially eliminates the Ni (0.007% max. wt. percent) from a Zircaloy 2-type alloy. Zircaloy 4 was developed as an improvement to Zircaloy 2 to reduce problems with hydriding, which causes Zircaloy 2 to become brittle when cooled to ambient temperatures (e.g. when the reactor is shut down) after absorbing hydrogen at higher temperatures.
The Zircaloy alloys are among the best corrosion resistant materials when tested in water at reactor operating temperatures (approx. 290.degree. C.) in the absence of the radiation from the nuclear fission reaction. The corrosion rate under those conditions is very low and the corrosion product is a uniform, tightly adherent, black ZrO.sub.2 film/layer. In actual service, however, the Zircaloy is irradiated and is also exposed to radiolysis products present in reactor water. The corrosion resistance properties of Zircaloy deteriorate under these conditions and the corrosion rate thereof is accelerated.
Research efforts directed at improving the corrosion properties of the zirconium-based alloys have yielded some advances. Corrosion resistance has been enhanced in some instances through carefully controlled heat treatments of the alloys either prior to or subsequent to material fabrication. Added heat treatment cycles, however, generally increase the expense of making finished products, and in those instances where an installation requires welding to be performed, the area affected by the heat of the welding operation may not possess the same corrosion resistance characteristics as the remainder of the article. Variations in the alloying elements employed and the percentages of the alloying elements have also been propounded in an effort to address corrosion-resistance deterioration of these alloys.
The deterioration under actual reactor conditions of the corrosion resistance properties of Zircaloy is not manifested in merely an increased uniform rate of corrosion. Rather, in addition to the black ZrO.sub.2 layer formed, a localized, or nodular corrosion phenomenon has been observed especially in boiling water reactors (BWR). In addition to producing an accelerated rate of corrosion, the corrosion product of the nodular corrosion reaction is a highly undesirable white ZrO.sub.2 bloom which is less adherent and lower in density than the black ZrO.sub.2 layer.
The increased rate of corrosion caused by the nodular corrosion reaction will be likely to shorten the service life of the tube cladding, and also this nodular corrosion will have a detrimental effect on the efficient operation of the reactor. The white ZrO.sub.2, being less adherent, may be prone to spalling or flaking away from the tube into the reactor water. On the other hand, if the nodular corrosion product does not spall away, a decrease in heat transfer efficiency through the tube into the water is created when the nodular corrosion proliferates and the less dense white ZrO.sub.2 covers all or a large portion of a tube.
Actual reactor conditions cannot be readily duplicated for normal laboratory research due to the impracticality of employing a radiation source to simulate the irradiation experienced in a reactor. Additionally, gaining data from actual use in reactor service is an extremely time consuming process. For this reason, there is no conclusory evidence in the prior art which explains the exact corrosion mechanism which produces the nodular corrosion. This limits, to some degree, the capability to ascertain whether other alloys will be susceptible to nodular corrosion before actually placing samples made from these alloys into reactors.
Laboratory tests conducted under the conditions normally experienced in a reactor (absent radiation) at approximately 300.degree. C. and 1000 psig in water, will not produce a nodular corrosion product on Zircaloy alloys like that found on Zircaloy alloys which have been used in reactor service. However, if steam is used, with the temperature increased to over 500.degree. C. and the pressure raised to 1500 psig, a nodular corrosion product like that found on Zircaloy in reactor service can be produced on zircaloy alloys in laboratory tests. Specimens of Zircaloy alloys which are annealed at 750.degree. C. for 48 hours are particularly susceptible to nodular corrosion under these test conditions. These annealed Zircaloy specimens will produce, in tests run for relatively short times, i.e. 24 hours, a degree of nodular corrosion comparable to that of Zircaloy tube cladding in actual reactor service. At this higher temperature and pressure, a simulated nuclear reactor environment is provided which will allow researchers to determine the susceptibility of new alloys to nodular corrosion. Results of these tests can be generally compared to those of Zircaloy specimens tested under the same conditions.
Any new alloy which would be considered as a suitable alternate or replacement for the Zircaloy alloys must not only be less susceptible than the Zircaloy alloys to nodular corrosion, but must maintain acceptable uniform corrosion rates, comparable to those of the Zircaloy alloys, to ensure sufficient service life.
It is therefore a principal object of the present invention to provide a group of alloys having improved corrosion resistance characteristics in a nuclear reactor environment.
It is another important object of the present invention to provide a group of alloys which do not depend on carefully controlled heat treatment for their corrosion resistance properties.